ML20207H944
| ML20207H944 | |
| Person / Time | |
|---|---|
| Issue date: | 08/22/1988 |
| From: | Butcher E Office of Nuclear Reactor Regulation |
| To: | Rossi C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8808300133 | |
| Download: ML20207H944 (8) | |
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AUG 2 21983
. MEMORANDUM FOR:
Charles E. Rossi, Director Division of Operational Events Assessment, NRR FROM:
Edward J. Butcher Chief Technical Specifications Branch Division of Operational Events Assessment, NRR
SUBJECT:
MINUTES OF MEETING WITH NUMARC STS IMPROVEMENT COORDINATION GROUP REGARDING PROGRESS OF WORK ON THE TECHNICAL SPECIFICATIONS IMPROVEMENT PROGRAM On July 28, 1988, the NRC staff met with the NUMARC STS Improvement Coordination Group regarding progress in developing new Standard Technical Specifications (STS).
NUMARC discussed its schedule and method of submitting new STS to the staff; they commented on technical specification "split" issues, key implementation issues, and the status of their proposed guidance document on 10 CFR 50.59.
The staff gave its current position on the technical specification "split" issues, the key implementation issues, and provided status of their work on line item improvements and surveillance testing.
Salient points of the meeting are mentioned below.
The staf f emphasized that it must be informed of any proposed technical changes to STS as soon as they are identified.
Early notification of such technical changes is necessary to support the staff's intended 6 to 9 month review of the new STS. Additionally the staff commented and industry acknowledged that, to date, industry has not proposed any changes to sections 5 or 6 of STS.
Regarding potential technical changes, NUMARC requested the staff to meet with the NUMARC working group for Reactivity Control systeais and Power Distribution Limits on August 10 to hear their presentation on proposed methods to develop new STS in the working group's area of responsibility.
The staff initiated a procedure by which the following would be documented at the end of each meeting between the staff and NUMARC:
The wording, responsible organization, and due date for action items resulting from the meeting. Accordingly, the lists of ACTION ITEMS developed during this meeting are enclosed.
The r. ext meeting between NRC and the NUMARC STS Improvement Coordination Group is schiduled for September 7, 1988, from 1:00 p.m. to 5:00 p.m.
Odr1 tS Signed By 8808300133 880822 Edward J. Butcher, Chief PDR REVGP ERGNUMRC Technical Specifications Branch PNU Division of Operational Events Assessment, NRR
Enclosures:
d (1) Action Items ih ff
[,f (2) Post Accident Monitoring Requirements in the Revised STS T
(3) Meeting Attendance i
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/g OlSTRIBUTION:
See attached fA:NRR0:00EA:NRR
\\' g TSB NRR TSB:DOEA:NRR C:T
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FMReinhart:pmc RLEmch EJ ut er 08/p/88 08/22/88 08/22/88 1 g
AUG 2 21988 DISTRIBUTION:
TEMurley VStello, EDO
<POR s" JHConran JHSniezek BBHayes OTSB Members CHBerlinger FJMiraglia MJClausen. OC OTSB R/F WDLanning TTMartin JLieberman. OE 00EA R/F WKennedy SAVarga WGMcDonald. 0 ARM Central Files DMCrutchfield WBKerr. SDBU/CR OTSB S/F - (MEETING NOTICE)
JGPartlow, OSP ELJordan AE00 Regional Administrators FJCongel PEBird FGillespie HLThompson, Jr. HMSS JWRoe NRC Participants SRConnelly, OlA WCParler. 0GC SJChilk. SECY BKGrimes LShao I
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ENCLO6URE (1)
OWNERS. GROUP ACTION ITDG ITD1 ORG.
DUE MTE
- 1. Review split report appendices A through D for and identify inconsistencies to NBC.
- 2. Develop proposal to control Bases.
- 3. OG's develop their proposed pre-noticing procees for license a umdments.
OG SEP 7, 1988 Meet with ctaff to discuss the acceptability of the OG's proposal.
plant specific split? Where do relocated requirements go? Come to the next meetire i
prepared to discuss these items.
I S. Update chapter euMittal schedule or inform UTSB that there are no changes.
-SEP 15, 1988 1
- 6. Review the PRAB guidance for evaluating PRA's for risk significance of Accident OG Upon Monitoring Instruments (Enclosed).
Reaching a Decision
- a. Decide n OG's want to take any action or not.
- b. Decide if OG's want to pursue changes to AUr's and Action Statements.
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ENCLO6URE (1)
STAFF ACTION ITIGE I'ID1 OIC.
DUE InTE
- 1. Send the NUMARC schedule for subnitting chapters to the technical branches.
OTSB AUG 10, 1988
- 2. Feedback to industry on sample Bases fran GE Topical Repor~ EtS-46-0487.
OTSC AUG 31, 1988 c
- 3. Decide if ? N>te Shutdown instrunentation may be relocated or if it must be retained UPSB SEP 7, 1988 in STS.
- 4. Remove note from Radiation Monitoring Instrumentatien which requires that all these OTSB SEP
- 7. 1983 instruments be retained as one system.
- 5. Review OG subnittaln and list the Radiation Monitoring Instrumentation in the OTSB SEP 7, 1988 following categories:
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- a. RETS.
- b. Operation Safety.
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- c. Other awlicable category (ies).
- 6. Make decision whether RETS are to be handled by Rulanaking or by Generic Letter.
DOEA SEP 7, 1988
- 7. Review industry's concem about wording of criteria clarification for ispect OTSB.SEP 7, 1988 of the SER and SRP on SAR. Decide if NRC w,rding of clarification needs revision.
- 8. Review industry's concem about wording of criteria clarification about transients OTSB SEP 7, 1988 and accidents outside of FSAR chapters 6 and 15.
Decide if NRC wording needs clarification.
9 Provide outline of review plan for STS and lead plant submir,tals.
OTSB SEP 7, 1988
- 7. 1988 process for license amendmento.
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PRAB Guidance to Owners Groups.
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2 6 MS POST ACCIDERT WNITORIm RFDUIREMERIS IN 'IME REVTun STANDARD TMICAI, SPECIFICATIONS
'lhe Technical Specification Policy Statement contains three criteria for detemining the retention of specific requiments within the Standard Technical Speci:tications (STS). With regard to Post Accident Monitoring Instruoentation (PAMI), the industry and the NRC staff agree that these criteria mquim that instrunents which provide infomation about Type A variables be retained in the STS. Type A variables an defined in Regulatory Gaide 1.97, Instrumentation for Light Water Reactors. They are variables which are needed by the operator in order to pefom actions tht are part of the primary success path of a design-basis accident.
In addition to the three criteria, h Policy Statement contains the requirement that the industry "verify that none of the requirements to be relo::ated contain constrainta of prime importance in limiting the likelihood or severity of the accident sequences that are creenly found to dominate risk". Among the requirementa proposed for relecatico are those Category I instruoents defined in Regulatory Guide 1.97 which am not Type A variables, h staff has raised a concern over their removal from the STS.
'Ihe operator uses a number of Category I instrumenta during accidet sequenoes which dominate risk. For example, he uses some to verify that autmatic safety functie< t occur (e.g., reactor scram, or auxiliary feedwater fics). He uses others to determine the need,to take manual corrective action (e.g.,
controlling rwnctor coolant pmesure or level). In view of this, the staff believes it is lihely that scoe of the Category I instruoenta in addition to Type A may be risk significant. Hence, the staff has zwquested that the industry provide further technical justificatico that relocating these instrumenta free the STS does not renove a constraint of prime importance to limiting plant risk.
In respcose to the staff's request for further technical justification, h industry requested the staff to provide guidance en the type of analysis that would be sufficient to addmss the st-ff's concerns. N purpose of this paper is to define the scope and limita e' ne proposed analysis. We r*==and thtd the industry consider the guidance provided below and sutnit a specific approach for our review, o % annivnia ahen1d be limited 'to Frident nacuanw and enerator actiona ubich have been wyielled stacifirmily in at.ata-ef-the-art PRAs.
Althcush the treatment of human errors in 11As has improved continuously over the past decah, intortant classes of operator actions still are not sodelled adequately. '1hese include cperator errors of==(==icn and operator recovery actions. In order to oover fully these types cf errors, the analysis wmld have to extend well beyond the cur mat limita ef PRA methodology.
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g PRAB Guidence to Ownero Groups.
h staff believes that it is the clear intent of the Policy Statement to cordine this study to sequences analysed in existing FRAs. Furthemore, we believe that most risk-significant instmmenta can be captumd by a study of a few well-modelled axisting PRAs.
i Hevertheless, the staff tweegnises that significant risk savings can be aoooeplished through enhanced operator moovery actions. Hence, as 4
part of the implementation of the Severs Accident Management program, 4
the staff has initiated a coopemtive yngram with the industry to improve Accident Management capabilities. It is likely that a pmgram of that type will identify additional variables which will help in rooovery actions, and the staff will be seeking assurances for those variables to be available in the event of an accident. However, the decisicn to add additional variables as a result of the Accident Management pregram to the list of technical specificaticms will have to be made at a time detemined by the program schedule, h current effort to define the scope of the STS should not await resolution of the Accident Management program, f
o N invest 4antian ahnm1r4 not be 14m4+m4 to Ca+manry I inntennanta.
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. h Category I classificatim of the instmmenta listed in Regulatory Guide 1.97 is well justified. It is based on a thorough censideration of plant operatien and accident management needs. However, just as we expect that some of them may be found to be of secondary risk significance, we also recognise that scoe non-Category I instrumenta may emerige from the industry study as being risk significant, hrifore, the study abould not be limited to the Category I list of instrumenta.
o h mussaa+m4 mathed of analvmins th v4 ak afsnifirmnam of elant innte=rita ia th avamina+ 4an of aaveral nlant swific PRAM.
bre are several plant specific PRAs which contain zwasonably detailed operator actica modeling. Some representative FRAs should be selected (e.g., Ooonse, N=Mam Neck, Limerick) and the leading eportter acticos identified on the basis of risk achievement importance or scoe similar measum. '!he key objective is to identify cases where a serious degradation in the error rate associated with operator action would significantly increase risk. For those leading operator acticms, the sensitivity to inadequate instrNeentation should be studied and factored into the c.etartinatice of risk significanoe.
In meeting the above objectives, the analysis should define and apply i
apropriate screening criteria to the leading operator actions oelected from PRAs. As indicated, the screening criteria should focus on operator errors which are sensiti';e to instumentation Jaalfuncticos.
Instrumenta selected on this basis would be candidates for retention within the Standart1 Technical Specificaties. As a minism, the criteria should address operstor erTors due to Watrument malfunctice, the need for multiple signals, and the availability and operability of altamate instrumentation.
Specifically, the criteria should be based on the following:
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,jul ! 6 El PRAB Guidenco to Ownsro Groups.
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- Is the information obtainable from altamate instnments?
- Is them adequate assurance that the altamate instnments am operable (ocaman mode failums am particularly significant)?
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h staff believes that the above approach would be sufficient to address the concems ngarding the development of the STS with respect to Category I as well as other instamentation. Other apprceches any be used, as long as they address the sensitivity of operator action errors to instrument degradation or malfunctice.
e he inat% m incinand in ahndard unhniem1 anarificationa ahmid be t
14mihA to thnam whteh mAAmma riak imanam of ranarie mieniflemnen.
For the purpose of defining instnmentation to be included in i
Standard Technical Specifications, industry analysis should be confined to sequenoes and operator actions which are judged to be of generlo sienificanoe.
We recognize that important plant specific accident sequenoes have been found in nearly all PRAs performed to date. However, the Individual Plant Examinaticms (IPEs), which will be perfomed under the Severs Accident Policy, are intended to identify and remedy problens identified within these sequenoes. To the extent that the remedies call for plant-specific poet accident monitoring improvements, they will be addressed on a plant-by-plant basis.
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Enclosure (3) b MEETINC ATTENDEES Name Affiliation R. L. Gill Duke Power C. W. Smyth GPU Nuclear R. A. Bernier Ariz. Public Service Co.
Stu Webster C-E Dan Foley C-E Tom Tipton NUMARC Les Bencini Westinghouse Tom Grozan Fla. Power & Light Co.
Samuel E. Bryan NRC/TSB Kulin 0. Desai NRC/TSB Millard Wohl NRC/TSB Mark Reinhart NRC/TSB Jim Klapproth GE R. R. Sgarro PP&L/BWROG R. A. Turner Babcock & Wilcox Art Bivens NUMARC N. J. Stringfellow Southern Co. Services Rich Emch NRC/TSB Oonald R. Hoffman BWR OG TSC I
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